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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20203F4511999-02-12012 February 1999 SER Finding Licensee Adequately Addressed GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves, for Virgil C Summer Nuclear Station ML20198F4241998-12-18018 December 1998 Safety Evaluation Granting Relief Request for Approval to Repair ASME Code Class 3 Service Water Piping Flaws in Accordance with GL 90-05 for VC Summer Nuclear Station ML20155G4551998-11-0404 November 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Use Code Case N-416-1 with Licensee Proposed Addl Exams ML20154Q9571998-10-21021 October 1998 SER Accepting Request Seeking Approval to Use Alternative Rules of ASME Code Case N-498-1 for Class 1,2 or 3 Sys ML20237A7181998-08-13013 August 1998 SER Accepting Util Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20248J0191998-06-0404 June 1998 Safety Evaluation Accepting Licensee Inservice Testing Program Interim Pump Relief Request Per 10CFR50.55a(a)(3) (II) ML20212C0381997-10-19019 October 1997 Safety Evaluation Accepting License Request for Deviation from Commitment to Meet Section III.G.2.c of App R to 10CFR50 Re Fire Protection of Safe Shutdown Capability for Plant ML20217E3491997-09-22022 September 1997 Safety Evaluation Authorizing Licensee Proposed Alternative for Current Interval Insp Program Plan ML20133J5551997-01-15015 January 1997 Safety Evaluation Granting Licensee Request Proposing Not to Perform Increased Frequency Testing on a Charging Pump at Virgil C Summer Nuclear Station ML20128G2931996-10-0202 October 1996 Safety Evaluation Supporting Amend 135 to License NPF-12 ML20128F4221993-02-0909 February 1993 Safety Evaluation Re Nuclear Physics Methodology for Reload Design.Request to Perform Reload Analyses Approved ML20056A7931990-08-0606 August 1990 Safety Evaluation Re Capability of Plant to Meet Requirements of Branch Technical Position Rsb 5-1, Design Requirements of RHR Sys. Design Satisfies License Condition 4 ML20245F5061989-06-22022 June 1989 Safety Evaluation Re Request for Relief from Section XI Re Hydrostatic Test Requirement ML20244D7361989-06-12012 June 1989 Safety Evaluation Accepting Util Response to Generic Ltr 83-28,Item 4.5.3 Re Reactor Trip Sys Reliability for All Domestic Operating Reactors ML20195B4421988-10-28028 October 1988 Safety Evaluation Supporting Amend 74 to License NPF-12 ML20151K0901988-07-28028 July 1988 Safety Evaluation Supporting Util Proposed Implementation of ATWS Rule Pending Resolution of Tech Spec Issue ML20151K7771988-07-27027 July 1988 Safety Evaluation Supporting Util Request to Deviate from Recommendations of Reg Guide 1.97 Re Instrumentation to Monitor Containment Temp ML20151R8561988-04-19019 April 1988 Safety Evaluation Supporting Related Inservice Testing Program & Request for Relief of Utils ML20236R4111987-11-13013 November 1987 Safety Evaluation Supporting Conformance to Reg Guide 1.97, Rev 3 ML20236K7701987-11-0505 November 1987 SER Accepting Util 831104 & 870401 Responses to Item 2.2.1 of Genreic Ltr 83-28 Re Equipment Classification Programs ML20237H3661987-07-22022 July 1987 Corrected Page to Safety Evaluation Issued W/Amend 67, Changing Second Paragraph & Deleting Third Paragraph on Page Three ML20214S8881987-06-0303 June 1987 Safety Evaluation Rept Granting Relief from Hydrostatic Testing After Repair to ASME Code Section Xi,Class 1,reactor Coolant Pump Seal Injection Line ML20213A5611987-01-30030 January 1987 SER Accepting Util 831104 Response to Generic Ltr 83-28,Item 2.1 (Part 2) Re Vendor Interface Program for Reactor Trip Sys Components ML20209H3331987-01-30030 January 1987 SER Supporting Util 831104 Response to Generic Ltr 83-28, Item 4.5.2 Re on-line Testing of Reactor Trip Sys Reliability ML20212F2841986-12-22022 December 1986 Safety Evaluation Supporting Amend 57 to License NPF-12 ML20211M4161986-12-0909 December 1986 Safety Evalution Supporting Licensee 860123 Submittals Re Fast Neutron Fluence for Fracture Toughness Requirements for Protection Against PTS Events (10CFR50.61) ML20203N0151986-09-15015 September 1986 Safety Evaluation Re Generic Ltr 83-28,Items 3.1.3 & 3.2.3, Post-Maint Testing (RTS Components,All Other Safety-Related Components). Response Acceptable ML20199D4211986-06-0909 June 1986 SER on Util 831104 & 860423 Responses to Generic Ltr 83-28, Item 1.2 Re post-trip Review Data & Info Capabilities.Data & Info Capabilities Acceptable ML20211A2571986-05-22022 May 1986 Safety Evaluation Accepting Mods to App R,Clarified by Generic Ltrs 81-12 & 83-33,to Prevent Spurious Equipment Operation Caused by fire-induced Conductor or Cable Faults, Facilitate Operator Actions & Resolve Addl Circuit Concerns ML20154A0621986-02-24024 February 1986 Safety Evaluation Supporting 850930 & 1204 Responses to 850802 & 1104 Requests,Respectively,For Addl Info Re Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events ML20154D1921986-02-14014 February 1986 Sser 1 Re Licensee 851204 Response to Generic Ltr 83-28, Item 3.2.2 Concerning Procedures & Programs to Review Info on safety-related Equipment.Response Acceptable & Meets Intent of Generic Ltr 83-28 ML20136B2291985-11-0707 November 1985 Safety Evaluation Supporting Amend 46 to License NPF-12 ML20209H8411985-11-0404 November 1985 Safety Evaluation Re Response to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2,4.1 & 4.5.1.Response to Item 3.2.2 Incomplete & Addl Info Required ML20137S5781985-09-24024 September 1985 SER Approving Licensee 831104 & 0715 Responses to Generic Ltr 83-28,Item 1.1 Re post-trip Review.Program & Procedures for Restart from Unscheduled Reactor Trip Acceptable ML20133H7321985-08-0202 August 1985 SER Re Util 831104 Response to Generic Ltr 83-28,Items 4.2.1 & 4.2.2 Concerning Reactor Trip Sys Reliability. Licensee Should Add Undervoltage Trip Attachment to Trending Program ML20128A2181985-06-21021 June 1985 SER of Util Response to Generic Ltr 83-28,Item 1.1 Re post-trip Review.Listed Addl Info Required Before Review Can Be Completed 1999-02-12
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217D6401999-10-31031 October 1999 Rev 2 to WCAP-15102, VC Summer Unit 1 Heatup & Cooldown Limit Curves for Normal Operation RC-99-0202, Monthly Operating Rept for Sept 1999 for VC Summer Nuclear Station.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for VC Summer Nuclear Station.With ML20216J4191999-09-24024 September 1999 Part 21 Rept Re 990806 Abb K-Line Breaker Defect After Repair.Vendor Notified of Shunt Trip Wiring Problem & Agreed to Modify Procedure for Refurbishment of Breakers RC-99-0180, Special Rept on 990807,electric Driven Fire Pump XPP0134A Was Declared Inoperable.Caused by Pump Discharge Relief Valve Failing to Open as Normally Expected.Two Temporary Fire Pumps Were Installed to Provide Backup Suppression1999-09-0808 September 1999 Special Rept on 990807,electric Driven Fire Pump XPP0134A Was Declared Inoperable.Caused by Pump Discharge Relief Valve Failing to Open as Normally Expected.Two Temporary Fire Pumps Were Installed to Provide Backup Suppression RC-99-0183, Monthly Operating Rept for Aug 1999 for Virgil C Summer Nuclear Station,Unit 1.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Virgil C Summer Nuclear Station,Unit 1.With ML20211K6161999-08-31031 August 1999 Rev 2 to VC Summer Nuclear Station,Colr for Cycle 12, Dtd Aug 1999 RC-99-0168, Special Rept:On 990804,electric Driven Fire Pump XPP0134A & Diesel Driven Fire Pump XPP0134B,were Removed from Svc to Allow for Plant Mod.Fire Pumps Were Returned to Operable Condition on 990818,after Mod Was Completed1999-08-19019 August 1999 Special Rept:On 990804,electric Driven Fire Pump XPP0134A & Diesel Driven Fire Pump XPP0134B,were Removed from Svc to Allow for Plant Mod.Fire Pumps Were Returned to Operable Condition on 990818,after Mod Was Completed ML20210M7071999-07-31031 July 1999 Rev 1 to VC Summer Nuclear Station COLR for Cycle 12 ML20211C2201999-07-31031 July 1999 Rev 1 to WCAP-15102, VC Summer Unit 1 Heatup & Cooldown Limit Curves for Normal Operation RC-99-0163, Monthly Operating Rept for July 1999 for VC Summer Nuclear Station,Unit 1.With1999-07-31031 July 1999 Monthly Operating Rept for July 1999 for VC Summer Nuclear Station,Unit 1.With RC-99-0137, Monthly Operating Rept for June 1999 for VC Summer Nuclear Station,Unit 1.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for VC Summer Nuclear Station,Unit 1.With RC-99-0122, Monthly Operating Rept for May 1999 for VC Summer Nuclear Station.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for VC Summer Nuclear Station.With ML20206H2971999-05-0505 May 1999 Part 21 Rept Re Common Mode Failure for magne-blast Breakers.Vc Summer Nuclear Station Utilizes These Breakers in Many Applications,Including 7.2-kV EDG Electrical Buses RC-99-0103, Monthly Operating Rept for Apr 1999 for VC Summer Nuclear Station.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for VC Summer Nuclear Station.With ML20206K2421999-04-30030 April 1999 Rev 0 to COLR for Cycle 12 for Summer Nuclear Station RC-99-0087, Part 21 Interim Rept (SSH 99-0001) Re 990218 Failure of Circuit Breaker Located in Cubicle 14 of XSW1DB to Close During Surveillance Testing.Caused by Positive Interlock Angle Was Incorrect.Breaker Was Returned to GE Factory1999-04-15015 April 1999 Part 21 Interim Rept (SSH 99-0001) Re 990218 Failure of Circuit Breaker Located in Cubicle 14 of XSW1DB to Close During Surveillance Testing.Caused by Positive Interlock Angle Was Incorrect.Breaker Was Returned to GE Factory RC-99-0083, Monthly Operating Rept for Mar 1999 for VC Summer Nuclear Station,Unit 1.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for VC Summer Nuclear Station,Unit 1.With RC-99-0063, Special Rept:On 990302 & 16,meteorological Site Number One 10 Meter Temp Element Was Declared Inoperable.Caused by Erratic Operation.Cabling & 10 Meter Electrical Connectors Were Replaced1999-03-26026 March 1999 Special Rept:On 990302 & 16,meteorological Site Number One 10 Meter Temp Element Was Declared Inoperable.Caused by Erratic Operation.Cabling & 10 Meter Electrical Connectors Were Replaced ML20196K5421999-03-22022 March 1999 Rev 2 to VC Summer Nuclear Station,Training Simulator Quadrennial Certification Rept,1996-99, Books 1 & 2. Page 2 of 2 Section 2.4.4 (Rev 2) of Incoming Submittal Were Not Included RC-99-0055, Special Rept:On 990302,Meteorological Site Number One 10 Meter Temp Element (RTD) Was Declared Inoperable Due to Erratic Operation.Cause of Original RTD Failure Is Unknown. Equipment Was Declared Operable on 9903121999-03-16016 March 1999 Special Rept:On 990302,Meteorological Site Number One 10 Meter Temp Element (RTD) Was Declared Inoperable Due to Erratic Operation.Cause of Original RTD Failure Is Unknown. Equipment Was Declared Operable on 990312 RC-99-0050, Monthly Operating Rept for Feb 1999 for VC Summer Nuclear Station,Units 1.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for VC Summer Nuclear Station,Units 1.With ML18106B0931999-02-25025 February 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Caused by Crack Due to Improper Location of Heated Bar.Only One Part Out of 7396 Pieces in Forging Lot Was Found to Be Cracked.Affected Util,Notified ML20203F4511999-02-12012 February 1999 SER Finding Licensee Adequately Addressed GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves, for Virgil C Summer Nuclear Station ML18106B0551999-02-0101 February 1999 Part 21 Rept Re Possible Matl Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Defect Is Crack in Center of Forging.Analysis of Part Is Continuing & Further Details Will Be Provided IAW Ncr Timetables.Drawing of Part,Encl ML18106B0441999-01-29029 January 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee Part Number SS-6-T.Caused by Crack in Center of Forging. Continuing Analysis of Part & Will Provide Details in Acoordance with NRC Timetables ML20206R5241998-12-31031 December 1998 Santee Cooper 1998 Annual Rept RC-99-0052, Vsns 1998 Annual Operating Rept. with1998-12-31031 December 1998 Vsns 1998 Annual Operating Rept. with RC-99-0004, Monthly Operating Rept for Dec 1998 for VC Summer Nuclear Station,Unit 1.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for VC Summer Nuclear Station,Unit 1.With ML20206R5191998-12-31031 December 1998 Scana Corp 1998 Annual Rept ML20198F4241998-12-18018 December 1998 Safety Evaluation Granting Relief Request for Approval to Repair ASME Code Class 3 Service Water Piping Flaws in Accordance with GL 90-05 for VC Summer Nuclear Station RC-98-0223, Special Rept 98-001:on 981130,steam Line High Range Gamma Monitor (RMG-19C) Was Declared Inoperable Due to Indeterminate Alarm.Caused by Failures in Detector & Meter Reset Circuitry.Established Preplanned Alternate Method1998-12-16016 December 1998 Special Rept 98-001:on 981130,steam Line High Range Gamma Monitor (RMG-19C) Was Declared Inoperable Due to Indeterminate Alarm.Caused by Failures in Detector & Meter Reset Circuitry.Established Preplanned Alternate Method RC-98-0222, Monthly Operating Rept for Nov 1998 for VC Summer Nuclear Station,Unit 1.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for VC Summer Nuclear Station,Unit 1.With ML20155G4551998-11-0404 November 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Use Code Case N-416-1 with Licensee Proposed Addl Exams RC-98-0208, Monthly Operating Rept for Oct 1998 for VC Summer Nuclear Station,Unit 1.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for VC Summer Nuclear Station,Unit 1.With ML20207J5701998-10-31031 October 1998 Non-proprietary Rev 1 to WCAP-14955, Probabilistic & Economic Evaluation of Rv Closure Head Penetration Integrity for VC Summer Nuclear Plant ML20154Q9571998-10-21021 October 1998 SER Accepting Request Seeking Approval to Use Alternative Rules of ASME Code Case N-498-1 for Class 1,2 or 3 Sys ML20154K7901998-09-30030 September 1998 Non-proprietary Rev 0 to WCAP-15101, Analysis of Capsule W from Sceg VC Summer Unit 1 Rv Radiation Surveillance Program RC-98-0184, Monthly Operating Rept for Sept 1998 for VC Summer Nuclear Station.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for VC Summer Nuclear Station.With ML20154K8041998-09-30030 September 1998 Non-proprietary Rev 0 to WCAP-15103, Evaluation of Pressurized Thermal Shock for VC Summer Unit 1 RC-98-0166, Monthly Operating Rept for Aug 1998 for VC Summer Nuclear Station,Unit 1.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for VC Summer Nuclear Station,Unit 1.With ML20237A7181998-08-13013 August 1998 SER Accepting Util Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves RC-98-0153, Monthly Operating Rept for July 1998 for VC Summer Nuclear Station,Unit 11998-07-31031 July 1998 Monthly Operating Rept for July 1998 for VC Summer Nuclear Station,Unit 1 RC-98-0131, Monthly Operating Rept for June 1998 for VC Summer Nuclear Station1998-06-30030 June 1998 Monthly Operating Rept for June 1998 for VC Summer Nuclear Station ML20248J0191998-06-0404 June 1998 Safety Evaluation Accepting Licensee Inservice Testing Program Interim Pump Relief Request Per 10CFR50.55a(a)(3) (II) RC-98-0113, Monthly Operating Rept for May 1998 for VC Summer Nuclear Station,Unit 11998-05-31031 May 1998 Monthly Operating Rept for May 1998 for VC Summer Nuclear Station,Unit 1 RC-98-0100, Monthly Operating Rept for Apr 1998 for VC Summer Nuclear Station,Unit 11998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for VC Summer Nuclear Station,Unit 1 ML20217G7411998-04-22022 April 1998 Rev 1 to VC Summer Nuclear Station COLR for Cycle 11 RC-98-0076, Final Part 21 Rept Re a DG EG-B for Vsns,As Followup to .Power Control Svcs of Engine Sys,Inc Provided Response on 980318.Evaluation Concludes That Failure of EG-B Is one-time non-repeatable Failure1998-04-17017 April 1998 Final Part 21 Rept Re a DG EG-B for Vsns,As Followup to .Power Control Svcs of Engine Sys,Inc Provided Response on 980318.Evaluation Concludes That Failure of EG-B Is one-time non-repeatable Failure RC-98-0084, Monthly Operating Rept for Mar 1998 for Virgil C Summer Nuclear Station,Unit 11998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Virgil C Summer Nuclear Station,Unit 1 ML20212H1421998-03-0202 March 1998 Interim Part 21 Rept SSH 98-002 Re EG-B Unit That Was Sent to Power Control Svcs for Determination of Instability & Refurbishment of a Dg.Cause of Speed Oscillations Unknown. Completed Hot Bore Checks on Power Case 1999-09-08
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g nee 3 % UNITED STATES j
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f NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066H001 k . . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE RELIEF REQUEST TO USE ASME CODE CASE N-498-1 FOR PERIODIC HYDROSTATIC TESTING OF ASME CLASS 3 PIPING SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION DOCKET NUMBER 50-395
1.0 INTRODUCTION
The Code of Federal Regulations,10 CFR 50.55a, requires that inservice inspection (ISI) of American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 systems be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda, except wMre the Commission grants specific written relief pursuant to 10 CFR 50.55a(g)(6)(i).10 CF, 50.55a(a)(3) indicates that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a cornpensating increase in the level of quality and safety.
Pursuant to 10 CFR 60.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements that become effective subsequent to editions specified in 10 CFR 50.55a(g)(2) and (g)(3), except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. Paragraph (g)(4)(ii) requires that inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals comply with the requirements of the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The second 10-year interval for the V. G. Summer Nuclear Station (VCSNS) began on January 1,1994, and ends on December 31,2003. The licensee's ISI program is based on the 1989 t dition of ASME Code Section XI. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modifications listed therein and subject to Commission approval.
9810260115 981021 PDR ADOCK 05000395 P PDR Enclosure
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Pursuant to 10 CFR 50.55a(g)(5)(iii), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is impractical for its facility, information shall be submitted to the Commission in support of that determination. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law; will not endanger life, property, or the common defense and security; and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
South Carolina Electric and Gas Company's (SCE&G's) July 1,1998, letter proposed an alternative examination to the requirements of the ASME Boiler and Pressure Code,Section XI.
SCE&G requested approval to implement the alternative rules of ASME Section XI Code Case N-498-1, dated May 11,1994, " Alternative Rules for 10-Year System Hydrostatic Testing for Class 1,2, and 3 Systems" pursuant to 10 CFR 50.55a(a)(3)(i) for 10-year hydrostatic testing on Class 1,2, and 3 systems. We have reviewed and evaluated the licensee's request and supporting information to use Code Case N-498-1 as a proposed alternative to the Code requirements for VCSNS.
2.0 EVALUATION 2.1 - Licensee's Request SCE&G requested relief from performing the Code-required hydrostatic test, and requested full implementation of Code Case N-498-1 for Class 1,2 and 3 systems.
2.1.1 Licensee's Component identification This request is for all ISI Class 1,2, or 3 system hydrostatic tests during each remaining ISI inspection interval.
2.1.2 ASME Code,Section XI, RequirementsSection XI, Table IWB-2500-1, Categories B-P, Table IWC-2500-1, Categories C-H, and IWD-2500-1, Categories D-A, D-B, and D-C contain the requirements for system hydrostatic and leakage testing. The Code requires system hydrostatic testing once per 10-year interval, at or near the end of the interval.
2.1.3 Licensee's Proposed Alternative Testing SCE&G proposed performing alternative examinations in accordance with ASME Code Case N-498-1, a system leakage test, in lieu of hydrostatic testing as follows:
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A system pressure test will be conducted during each inspection interval of Inspection Progr&m B.
l System pressure testing will extend to all Class 3 components included in those i
portions of systems required to operate or support the safety system function up to and including the first normally closed valve, including a safety or relief valve, or i valve capable of automatic closure when the safety function is required.
The system will be pressurized to nominal operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems, and 10 minutes for non-insulated systems. The system will be maintained at nominal operating pressure during the performance of the VT-2
, visual examination.
l The VT-2 visual examination will include all co" >nents within the boundary 1 identified above.
2.1.4 Licensee's Basis for Relief i
The licensee's basis for relief is as follows:
Currently the ASME Code requires that a hydrostatic test be performed on all Class 1,2 and 3 systems once each 10-year interval as a part of the inservice inspection plan. Even at the higher hydrostatic test pressures, the contribution of the pressure component to the overall design loads is relatively small. Based on this, the ASME, industry, and NRC, have realized that the hydrostatic test is not intended to be a test of structuralintegrity but, rather, an enhanced leakage test.
The value of hydrostatic testing in determining structural integrity is negligible.
Industry experience indicates that leaks are not being discovered where the leak originated due to hydrostatic test pressures causing a pre-existing flaw to propagate through-wall. Instead, the industry experierce shows that the majority of all leaks discovered were leaks which originated at normal operating pressures. Therefore, compared to a hydrostatic test, a normal pressure ;
leakage test is equally effective for discovering through-wall flaws.
Currently, licensees incur the cost of considerable time, radiation exposure and dollar resources carrying out hydrostatic test requirements. A significant effort may be necessary, depending on the system or plant configuration, system Code Class and other factors, to temporarily remove or disable Code safety and/or relief valves to meet test pressure requi*ements. The safety assurance provided by the enhanced leakage gained from a slight increase in system pressure during a hydrostatic test is offset by having to gag or remove Code safety and/or relief valves, placing the system in an off-normal stata, erecting
temporary supports in steam lines, possible extens!on of refueling outagas and recource requirements to set up testincj with special equipment and gages.
2.1.5 Evaluation 1' formation prepared in conjunction with ASME Code Case N-498-1 notes that the system hydrostatic test is not solely a test of system structuralintegrity, but also provides a means to enhance leakage detection. A paper by S. H. Bush and R. R. Maccary, "Cevelopment of /n-Service Inspection Safety Philosophy for U.S.A. Nuclear Power Plants, " ASME,1971, indicated that this was the originalintent. Pipirg components are designed for a number of loadings that would be postulated to occur under the various modes of plant operation. Hydrostatic testing only subjects the pipirig components to a smallincrease in pressure over the design pressure.
It does not present a significant challenge to pressure boundary integrity. Piping dead weght, thermal expansion and seismic loads, which may present far greater challenge to the structural ;
integrity of a system than fluid pressure, are not part of the loading imposed during a l hydrostatic test. Wateris used as a test medium in the hydrostatic test. Because wateris !
highly hcompressible, any small leak from a high-pressurized water-solid system can be readily detected by a sharp decline in system pressure, or by continual pumping required to maintain I the system pressure. As such, hydrostatic pressure testing is primarily regarded as a means to enhance leakage detection during the examination of components under pressure. Hydrostatic pressure testing provides good indication of any system leakages, especially those that might originate from small through-wall cracks of the pressure boundary. Consequently, this Code-required in-service hydrostatic pressure test enhances the possibility of timely discovery of '
small through-wal; flaws which, because of a tiny leak size, might not be readily detected by any other means such as system walkdowns or installed leak-detaction systems. !
SCE&G requested approval to implement the alternative rules of ASME Section XI Code Case N-498-1, in lieu of 10-year hydrostatic testing of Class 1,2, and 3 systems. The I;censee may already use N-498, " Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, and 2 Systems," since NRC Regulatory Guide 1.147, Rev.11 approves the use of Code Case N-498 for Class 1 and 2 systems. The rules for Code Class 1 and 2 in N-498-1 are unchanged from those in N-498. The staff found N-498 acceptabie because the Lternative of performing a test at a system pressure that is slightly lower than the hydrostatic pressure provided adequate assurance, and because compliance with the specified requirements would resuit in hardship cr unusual difficulty without a compensating increase in the level of quality and safety.
Code Case N-498-1 was revised from N-498 to ancompass Class 3 components, and specif;es requirements for Cless 3 that are identical to those for Class 2 components. In lieu of 10-year hydrostatic pressure testing at or near the end of the 10-year interval, Code Case N-498-1 requires a visual examination (VT-2) be performed in conjunctior, with a system leakage test in accordance with paragraph IWA-5000. A system leakage test may be conducted to derr.onstrate that leaks from pressure boundary that may originate from through-wall flaws do I
not exist. This would meet the intent of the hydrostatic test as noted above.
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Class 3 systems do not normally receive the amount or type of non-destructive examinations I that Class 1 and 2 systems receive. While Class 1 and 2 system failures are relatively uncommor,, Clast,3 system leaks occur more frequently, and the failure mode typically differs.
Based on a review of Class 3 system failures requiring repair for the last 5 years in Licensee Event Reports and the Nuclear Plant Reliability Data System databases, the most common causes of failures are erosion-corrosion (EC), microbiologically induced corrosion (MIC), and 9eneral corrosion. Licensees generally have programs in place for prevention, detection, and evaluation of EC and MIC. Leakage from general corrosion is readily apparent to inspectors when pedorming a VT-2 examination during system pressure tests.
l Since a hydrostatic test provides only a minimal amount of increased assurance over a system !
leakage test, and the . system pressure test provides adequate assurance, the staff finds that the proposed alternatives provide an acceptable level of quality and safety. Accordingly, the
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licensee's proposed elternative to use Code Case N-498-1 is authorized for VCSNS for the l
current interval, pursuant to 10 CFR 50.55a(a)(3)(i).
3.0 CONCLUSION
S l
, The staff evaluated SCE&G's information in support of its request for relief. Based on the i
informatica submitted, the alternative for hydrostatic testing contained in the licensee's proposal is authcrized pursuant to 10 CFR 50.55a(a)(3)(i) for Class 1,2, and 3 systems. The staff concludes that alternative rules of ASME Section XI Code Case N-498-1, in lieu of 10 year hydrostatic testing of Class 1,2, and 3 systems, provide an acceptable level of quality and safety. The SCE&G alternative is authorized for the duration of the currently approved ISI program plan, or until the Code Case is approved for general use by reference in l Regulatory Guide 1.147.
Principal Contributor: M. Padovan i i Date: October 21, 1998 l
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